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Power Engineering; v. 76(12); p. 30-37
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Yosifon, S.
Israel Atomic Energy Commission, Beersheba. Nuclear Research Center-Neguev1972
Israel Atomic Energy Commission, Beersheba. Nuclear Research Center-Neguev1972
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No abstract available
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Jan 1972; 32 p
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Report
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Dorr, B.; Huber, F.; Menzenhauer, P.; Peppler, W.; Till, W.
Reactor meeting, Karlsruhe 10.4.-13.4.19731973
Reactor meeting, Karlsruhe 10.4.-13.4.19731973
AbstractAbstract
No abstract available
Original Title
Das Brandverhalten von fluessigem Natrium
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Deutsches Atomforum e.V., Bonn (F.R. Germany); p. 75-78; 1973; ZAED; Leopoldshafen, F.R. Germany; Reactor meeting; Karlsruhe, F.R. Germany; 10 Apr 1973; 2 figs. Short communication only.
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Neuberger, M.; Bletzer, P.; Fronhoefer, M.; Hardy, T.; Kuhn, W.; Ricketts, C.I.; Schmidmeier, K.H.
Safety oriented LWR research. Annual report 19901991
Safety oriented LWR research. Annual report 19901991
AbstractAbstract
[en] To help estimate the mechanical loading at the service location two fluid dynamic codes were investigated. FLOWMASTER shows good results in generating the air-cleaning network and in calculating steady state conditions, but fails in calculating fluid dynamic transients. PROMO is very time consuming in generating the network, but one can obtain good results in calculating fluid dynamic transients. In addition to the challenges posed by elevated pressure, temperature, and air flow, the effects of shock waves, caused by explosions or hydrogen detonations must be considered. The problem of a 'shock-shock' in the flow after the diverging channel by high shock strengths were solved. In order to guarantee the safety margins of filter units during their entire service lives, filter performance needs to be verified under standardized test conditions that take into consideration the particularly adverse effects of filter exposure to super-saturated airflows and elevated differential pressure. The detail construction for a rig to type test filter units under fog conditions and increased pressure drop was finished, and the components of the test facility were chosen. (orig./DG)
Original Title
Stoerfallbeanspruchung innerhalb von Lueftungsanlagen
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Kernforschungszentrum Karlsruhe GmbH (Germany). Projekt Nukleare Sicherheitsforschung; 270 p; Jul 1991; p. 131-139
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Report
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AbstractAbstract
[en] A systematic approach is developed for the prediction of the performance of single, as well as double-entry, centrifugal pumps in a specific speed range of 600 to 2500 engineering units. Complete pump characteristics can be mapped, from shut-off conditions to 130% of the design flowrate. Given the dimensions of a centrifugal pump, as well as the impeller rotational speed, the method can be employed in the prediction of head vs flowrate (H-Q), shaft power vs flowrate (BHP-Q) and efficiency vs flowrate (E-Q) curves. Pump characteristics can then be used to model pump performance in state-space. This allows for development of optimal control strategies, for the design of controllers, and for examination of the sensitivity of operation to changes in design and performance parameters. Hence, optimization of process design and operation becomes attainable. The application of the method to a nuclear-grade coolant pump demon-strated the viability of the model as a tool in simulation of pump response to various inputs. State-space representation is shown to be invaluable in the simulation of performance at normal and abnormal operation conditions, synthesis of surveillance schemes and in detection of abnormalities caused by degradation of the components. (author)
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AbstractAbstract
[en] Safety valve rules, i.e., rules for overpressure protection by the use of various pressure relieving devices, vary somewhat among the five book sections of the ASME Boiler and Pressure Vessel Code which require such protection. This paper reviews those rules by discussing the following topics: Pressure relief device terminology and function. The problem of overpressure protection. Code rules for overpressure protection: rules for determining required relieving capacity; for allowable overpressure, for set pressure and set pressure tolerance; for blowdown. The various pressure relief devices permitted by the Code. Design of pressure relief valves. How relieving capacities are established and certified. The qualification of pressure relief device manufacturers. Installation guidelines. Concluding remarks. (orig.)
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Vereinigung der Technischen Ueberwachungsvereine e.V., Essen (Germany); 857 p; 1992; p. 151-173; 7. international conference on pressure vessel technology (ICPVT-7); 7. Internationale Konferenz ueber Druckbehaeltertechnologie; Duesseldorf (Germany); 31 May - 5 Jun 1992; Available from FIZ Karlsruhe
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AbstractAbstract
[en] By the nature of its design, the set point and lift of a conventional spring loaded safety relief valve are sensitive to back pressure. One way to reduce the adverse effects of the back pressure on the safety relief valve function is to install a balanced bellows in a safety relief valve. The metallic bellows has a rather wide range of manufacturing tolerance which makes the design of the bellows safety relief valve very complicated. The state-of-the-art balanced bellows safety relief valve can only substantially minimize, but cannot totally eliminate the back pressure effects on its set point and relieving capacity. Set point change is a linear function of the back pressure to the set pressure ratio. Depending on the valve design, the set point correction factor can be either greater or smaller than unity. There exists an allowable back pressure and critical back pressure for each safety relief valve. When total back pressure exceeds the Ra, the relieving capacity will be reduced mainly resulting from the valve lift being reduced by the back pressure and the capacity reduction factor should be applied in valve sizing. Once the Rc is exceeded, the safety relief valve becomes unstable and loses its over pressure protection capability. The capacity reduction factor is a function of system overpressure, but their relationship is non-linear in nature. (orig.)
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Source
Vereinigung der Technischen Ueberwachungsvereine e.V., Essen (Germany); 857 p; 1992; p. 174-182; 7. international conference on pressure vessel technology (ICPVT-7); 7. Internationale Konferenz ueber Druckbehaeltertechnologie; Duesseldorf (Germany); 31 May - 5 Jun 1992; Available from FIZ Karlsruhe
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AbstractAbstract
[en] This communication discusses the worldwide state of progress in the decommissioning of nuclear facilities. In general the technology of decommissioning is straightforward, but although there are many examples of successful decommissioning projects the general progress is slow. The reasons for this are identified as primarily economic, but the lack of waste disposal facilities and the unnecessary personnel radiation doses that may be incurred by early decommissioning are also factors. Finally the redundant structures are relatively safe and stable and there is little technical incentive to dismantle them. The current delays are heavily influenced by political factors which may be subject to change. (author)
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AbstractAbstract
[en] This standard is to be applied to ventilation components of ventilation classes 1 and 2 (see DIN 25 414) for nuclear power stations and similar components in nuclear plants in as far as they are important for radiation protection. (orig.)
[de]
Diese Norm ist anzuwenden fuer lueftungstechnische Komponenten, die in Kernkraftwerken zu den Lueftungsklassen 1 und 2 (siehe DIN 25 414) gehoeren, und fuer vergleichbare Komponenten in anderen kerntechnischen Anlagen, soweit diese Komponenten eine strahlenschutztechnische Bedeutung haben. (orig.)Original Title
Lueftungstechnische Komponenten in kerntechnischen Anlagen
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May 1992; 7 p; Beuth; Berlin (Germany); DIN--25496
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Book
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Standard
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Bernsteiner, F.J.; Muller, H.
Proceedings of the national symposium on advances in utility systems for industrial and nuclear installations (held at Bombay during January 9-11, 1992)1992
Proceedings of the national symposium on advances in utility systems for industrial and nuclear installations (held at Bombay during January 9-11, 1992)1992
AbstractAbstract
[en] The occurrence of several minor and an insignificant number of major cracks in shafts of main coolant and boiler feed pupms found during normal inspections in recent years was a challenge for launching extensive investigations to discover the causes and to seek appropriate remedial measures. Comprehensive experimental tests and analytical studies such as fatigue strength tests, measurement of residual stresses with electronic methods, finite element and crack origination and propagation calculations were carried out. The result reveal noticeable influence of residual stresses set up in the fabrication processes as well as the material surface coating and the water chemistry on fatigue strength behaviour of shafts in conjuction with the other applied stresses. Pump shaft manufactured by improved technologies and using a spline faced connection between shaft and impeller can withstand cyclic loads for the whole lifetime. Due to the very low vibration levels, safe and simple assembly and advantages for replacement of parts, some plants are converted to the new reactor design. (author). 11 figs
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Department of Atomic Energy, Bombay (India). Board of Research in Nuclear Sciences; 825 p; Jan 1992; p. 7.1.1-7.1.19; Bhabha Atomic Research Centre; Bombay (India); National symposium on advances in utility systems for industrial and nuclear installations; Bombay (India); 9-11 Jan 1992
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